NRC Generic Letter 1990-002: Alternative Requirement for Fuel Assemblies in Design Features Section of Technical SpecificationsML031140555 |
Person / Time |
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Site: |
Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, 05000000, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Clinch River |
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Issue date: |
02/01/1990 |
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From: |
Partlow J Office of Nuclear Reactor Regulation |
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To: |
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References |
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GL-90-002, NUDOCS 9001290051 |
Download: ML031140555 (6) |
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Similar Documents at Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, 05000000, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Clinch River |
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Category:NRC Generic Letter
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(Errata): Laboratory Testing of Nuclear-Grade Activated Charcoal NRC Generic Letter 1983-111999-06-24024 June 1999 NRC Generic Letter 1983-011, Supplement 1: Licensee Qualification for Performing Safety Analysis ML0823509351999-06-0303 June 1999 Generic Ltr 99-02 to All Holders of OLs for Nuclear Power Reactors,Except Those Who Have Permanenetly Ceased Operations & Certified That Fuel Permanently Removed from Rv Re Laboratory Testing of nuclear-grade Activated Charcoal ML0311101371999-06-0303 June 1999 Withdrawn NRC Administrative Letter 1999-002: Operating Reactor Licensing Action Estimates NRC Generic Letter 1999-011999-05-0303 May 1999 NRC Generic Letter 1999-001: Recent Nuclear Material Safety and Safeguards Decision on Bundling Exempt Quantities NRC Generic Letter 1998-011999-01-14014 January 1999 NRC Generic Letter 1998-001, Supplement 1: Year 2000 Readiness of Computer Systems at Nuclear Power Plants ML0311101601998-08-0303 August 1998 Withdrawn NRC Administrative Letter 1998-005: Availability of Summaries in Electronic Format of Technical Reports by Office for Analysis & Evaluation of Operational Data NRC Generic Letter 1998-041998-07-14014 July 1998 NRC Generic Letter 1998-004: Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-of-coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material i NRC Generic Letter 1998-031998-06-22022 June 1998 NRC Generic Letter 1998-003; NMSS Licensees and Certificate Holders Year 2000 Readiness Programs NRC Generic Letter 1998-021998-05-28028 May 1998 NRC Generic Letter 1998-002: Loss of Reactor Coolant Inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition NRC Generic Letter 1997-061997-12-30030 December 1997 NRC Generic Letter 1997-006: Degradation of Steam Generator Internals NRC Generic Letter 1997-051997-12-17017 December 1997 NRC Generic Letter 1997-005: Steam Generator Tube Inspection Techniques NRC Generic Letter 1996-061997-11-13013 November 1997 NRC Generic Letter 1996-006, Supplement 1: Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions ML0312007011997-10-0808 October 1997 Withdrawn NRC Generic Letter 1991-018, Revision 1: Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions NRC Generic Letter 1997-041997-09-30030 September 1997 NRC Generic Letter 1997-004: NRC Staff Approval for Changes to 10 CFR Part 50, Appendix H, Reactor Vessel Surveillance Specimen Withdrawal Schedules NRC Generic Letter 1997-031997-07-0909 July 1997 NRC Generic Letter 1997-003: Annual Financial Surety Update Requirements for Uranium Recovery Licensees NRC Generic Letter 1997-021997-05-15015 May 1997 NRC Generic Letter 1997-002: Revised Contents of Monthly Operating Report NRC Generic Letter 1995-061997-01-31031 January 1997 NRC Generic Letter 1995-006: Changes in Operator Licensing Program NRC Generic Letter 1996-081996-12-15015 December 1996 NRC Generic Letter 1996-008: Interim Guidance on Transportation of Steam Generators NRC Generic Letter 1996-051996-09-18018 September 1996 NRC Generic Letter 1996-005: Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves NRC Generic Letter 1996-041996-06-26026 June 1996 NRC Generic Letter 1996-004: Boraflex Degradation in Spent Fuel Pool Storage Racks NRC Generic Letter 1995-091996-04-0505 April 1996 NRC Generic Letter 1995-009: Supplement 1: Monitoring and Training of Shippers and Carriers of Radioactive Materials NRC Generic Letter 1996-021996-02-13013 February 1996 NRC Generic Letter 1996-002: Reconsideration of Nuclear Power Plant Security Requirements Associated with an Internal Threat NRC Generic Letter 1996-031996-01-31031 January 1996 NRC Generic Letter 1996-003: Relocation of the Pressure Temperature Limit Curves & Low Temperature Overpressure Protection System Limits NRC Generic Letter 1989-101996-01-24024 January 1996 NRC Generic Letter 1989-010, Supplement 7: Consideration of Valve Mispositioning in Pressurized-Water Reactors NRC Generic Letter 1996-011996-01-10010 January 1996 NRC Generic Letter 1996-001: Testing of Safety-Related Logic Circuits NRC Generic Letter 1995-101995-12-15015 December 1995 NRC Generic Letter 1995-010: Relocation of Selected Technical Specifications Requirements Related to Instrumentation ML0310701501995-10-31031 October 1995 Withdrawn - NRC Generic Letter 1995-008: 10 CFR 50.54(p) Process for Changes to Security Plans Without Prior NRC Approval NRC Generic Letter 1993-031995-10-20020 October 1995 NRC Generic Letter 1993-003: Verification of Plant Records NRC Generic Letter 1995-071995-08-17017 August 1995 NRC Generic Letter 1995-007: Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves NRC Generic Letter 1995-041995-04-28028 April 1995 NRC Generic Letter 1995-004: Final Disposition of the Systematic Evaluation Program Lesson-Learned Issues NRC Generic Letter 1995-031995-04-28028 April 1995 NRC Generic Letter 1995-003: Circumferential Cracking of Steam Generator Tubes NRC Generic Letter 1995-021995-04-26026 April 1995 NRC Generic Letter 1995-002: Use of Numarc/Epri Report TR-102348, Guideline on Licensing Digital Upgrades, in Determining the Acceptability of Performing Analog-To-Digital Replacements Under 10CFR 50.59 NRC Generic Letter 1995-011995-01-26026 January 1995 NRC Generic Letter 1995-001: NRC Staff Technical Position on Fire Protection for Fuel Cycle Facilities ML0312004431994-09-0202 September 1994 Withdrawn - NRC Generic Letter 1994-004: Voluntary Reporting of Additional Occupational Radiation Exposure Data NRC Generic Letter 1994-031994-07-25025 July 1994 NRC Generic Letter 1994-003: Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors NRC Generic Letter 1994-021994-07-11011 July 1994 NRC Generic Letter 1994-002: Long-Item Solutions and Upgrade of Interim Operating Recommendations for Thermal Hydraulic Instabilities in Boiling Water Reactors NRC Generic Letter 1994-011994-05-31031 May 1994 NRC Generic Letter 1994-001: Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators NRC Generic Letter 1993-081993-12-29029 December 1993 NRC Generic Letter 1993-008: Relocation of Technical Specification Tables of Instrument Response Time Limits NRC Generic Letter 1993-071993-12-28028 December 1993 NRC Generic Letter 1993-007: Modification of Technical Specification Administrative Control Requirements for Emergency & Security Plans NRC Generic Letter 1993-061993-10-25025 October 1993 NRC Generic Letter 1993-006: Research Results on Generic Safety Issue 106, Piping and the Use of Highly Combustible Gases in Vital Areas. NRC Generic Letter 1993-051993-09-27027 September 1993 NRC Generic Letter 1993-005: Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements for Testing During Power Operation NRC Generic Letter 1993-041993-06-21021 June 1993 NRC Generic Letter 1993-004: Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies, 10 CFR 50.54(f) NRC Generic Letter 1993-021993-03-23023 March 1993 NRC Generic Letter 1993-002: Public Workshop on Commercial Grade Procurement and Dedication NRC Generic Letter 1993-011993-03-0303 March 1993 NRC Generic Letter 1993-001: Emergency Response Data System Test Program 2023-08-03 The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:05000000]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] OR [[:Clinch River]] </code>.
[Table view]The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:05000000]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] OR [[:Clinch River]] </code>. |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, D. C. 20555 February 1, 1990
TO ALL LIGHT-WATER REACTOR LICENSEES AND APPLICANTS
SUBJECT: ALTERNATIVE REQUIREMENTS FOR FUEL ASSEMBLIES IN THE DESIGN FEATURES
SECTION OF TECHNICAL SPECIFICATIONS (Generic Letter 90-02)
Technical Specifications (TS) include design requirements for the fuel assemblies in Section 5, "Design Features." On a plant-specific basis, staff has approved changes to these requirements that provide flexibility for improved to fuel performance by permitting timely removal of fuel rods that are found sources be leaking during a refueling outage or are determined to be probable of future leakage. Because these improvements in a licensee's fuel performance program will provide for reductions in future occupational radiation exposure and plant radiological releases, this alternative is being made available to all plants as a line-item improvement in TS.
The requirements included in Section 5 of the TS on "Design Features" geometricaddress those features of the facility such as materials of construction and arrangements which, if altered or modified, would have a significant effect on safety and are not covered under other sections of the TS on "Safety Limits,"
"Limiting Conditions for Operation," or "Surveillance Requirements."forRecently, "Fuel the staff approved a lead-plant request to modify the requirements Assemblies." Specification 5.3.1, for the McGuire and Catawba plants. That rods or change allows the substitution of Zircaloy-4 or stainless steel filler cycle-specific open water channels for fuel rods in fuel assemblies if justified by change reload analyses using an NRC-approved methodology. In addition, the requires that a special report describing the number of rods replaced assembly, be submitted to NRC if more than 30 rods in the core, or 10 rods in any are replaced per refueling.
The enclosed guidance addresses the preparation of a license amendment request for this change to TS. Licensees and applicants are encouraged to
900129005
-4 propose changes to their TS that are consistent with the guidance in the Conforming amendment requests will be expeditiously reviewed by the NRC enclosure.
Manager for the facility. Please contact the Project Manager if you haveProject questions on this matter.
Sincerely, J es G. Partlow Associate Director for Projects Office of Nuclear Reactor Regulation Enclosures:
1. Guidance for the Modification of the Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications
2. List of Most Recently Issued NRC
Generic Letters eRH
i..
Generic Letter 9 0-.Q? Enclosure GUIDANCE FOR THE MODIFICATION OF THE REQUIREMENTS FOR FUEL ASSEMBLIES
IN THE DESIGN FEATURES SECTION OF TECHNICAL SPECIFICATIONS
INTRODUCTION
This enclosure provides guidance for the preparation of a license amendment request to modify the Technical Specifications (TS) requirements for "Fuel Assemblies" in the Section 5, "Design Features," of TS. This change to the TS
permits the substitution of Zircaloy-4 or stainless steel filler rods or open water channels for fuel rods in fuel assemblies if justified by cycle-specific reload analyses.
DISCUSSION
The requirements for fuel assemblies specify the quantity of fuel assemblies and the number of fuel rods per assembly. Flexibility to deviate from the number of fuel rods per assembly is desirable to permit timely removal of fuel rods that are found to be leaking during a refueling outage or are determined to be probable sources of future leakage. This improvement in the licensee's fuel performance program will provide for reductions in future occupational radiation exposure and plant radiological releases.
The substitution of filler rods or open water channels for fuel rods is accept- able when justified by cycle-specific reload analyses using an NRC-approved methodology. This reload analyses will demonstrate that existing design limits and safety analyses criteria are met in advance of the next operating cycle.
An "NRC-approved methodology" includes those methodologies acknowledged in the Final Safety Analysis Report and applied in support of the issuance of the original operating license. It also includes those subsequent methodologies that have been submitted to and accepted by the NRC staff as amendments to the operating license.
If the reconstitution of fuel assemblies through the use of filler rods or open water channels is extensive, this information should be reported to NRC.
Therefore, if more than 30 rods in the core, or 10 rods in any assembly, are replaced during a refueling, a special report describing the number of rods replaced must be submitted to NRC in accordance with the provisions of the TS
for special reports.
The attached example is a typical specification for fuel assemblies with the required modifications shown underlined. Proposed changes to plant TS should follow this guidance.
SUMMARY
The modification of the "Design Features" section of the TS on "Fuel Assemblies"
will not result in modifications to fuel assemblies that would have a signifi- cant effect on safety because of the necessity to justify such changes using an NRC-approved methodology. This requirement will confirm conformance to existing design limits and that safety analyses criteria are met before oper- ation during the next fuel cycle. A license amendment request to modify the requirements for "Fuel Assemblies" that is consistent with the guidance provid- ed will result in flexibility for improved fuel performance.
Generic Letter 90-02 Enclosure MODEL TECHNICAL SPECIFICATION CHANGE
(CHANGES SHOWN UNDERLINED)
5.0 DESIGN FEATURES
5.3 REACTOR CORE
FUEL ASSEMBLIES
5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly nominally containing 264 fuel rods clad with Zircaloy-4, 3except that substitu- tibini577ircaloy-4 or stainless steel filler rods or open water channels for fuel rods may be ma-de in fuel assemblies if justified by cycle-speciffic Rreoad analyses using an NRC-a-pproved methodolo y. Should more than 30 rods in-the core. or 10 rods in any assembly be replaced per-refuel ing a secialreo~rt describing tenumber o rodsreplaced shal besubmited to te Commission pursuant to Specification 6.9.2 within 30-days after c cle startup Each fuel rod shal have a nomina active fuel lengt of 14 nches. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 4.0 weight percent U-235.
Specification 6.9.2 of the Standard Technical Specifications (STS) states:
Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.
Plants whose Technical Specifications do not include STS 6.9.2 should ensure that the revised Specification 5.3.1 includes reporting requirements equivalent to the combination of STS 5.3.1 and 6.9.2 above.
A..
LIST OF RECENTLY ISSUED GENERIC LETTERS
Generic Date of Letter No. Subject Issuance Issued To Issued To
90-01 REQUEST FOR VOLUNTARY 01/18/90 ALL LICENSEES OF
PARTICIPATION IN NRC OPERATING REACTORS &
REGULATORY IMPAC SURVEY CONSTRUCTION PERMITS
FOR LWR NUCLEAR POWER
PLANTS
89-23 NRC STAFF RESPONSES TO 10/23/89 ALL HOLDERS OF
QUESTIONS PERTAINING TO OPERATING LICENSEES
IMPLEMENTATION OF 10 CFR AND CONSTRUCTION
PART 26 - GENERIC LETTER PERMITS FOR NUCLEAR
89-23 POWER PLANTS
89-22 POTENTIAL FOR INCREASED ROOF 10/19/89 ALL LICENSEES OF
LOADS AND PLANT AREA FLOOD OPERATING REACTORS AND
RUNOFF DEPTH AT LICENSED HOLDERS OF CONSTRUCTION
NUCLEAR POWER PLANTS DUE TO PERMITS (EXCEPT BYRON
RECENT CHANGE IN PROBABLE BRAIDWOOD, VOGTLE,
MAXIMUM PRECIPITATION SOUTH TEXAS, AND RIVER
CRITERIA DEVELOPED BY THE BEND)
NATIONAL WEATHER SERVICE
(GENERIC LETTER 89-22)
89-21 REQUEST FOR INFORMATION 10/19/89 ALL HOLDERS OF
CONCERNING STATUS OF OPERATING LICENSES
IMPLEMENTATION OF UNRESOLVED AND CONSTRUCTION
SAFETY ISSUE (USI) REQUIREMENTS PERMITS FOR NUCLEAR
POWER REACTORS
89-20 PROTECTED AREA LONG-TERM 09/26/89 ALL FUEL CYCLE
HOUSEKEEPING FACILITY LICENSEES
WHO POSSESS, USE,
OR PROCESS FORMULA
QUANTITIES OF
STRATEGIC SPECIAL
NUCLEAR MATERIAL
89-19 REQUEST FOR ACTION RELATED TO 09/20/89 ALL LICENSEES OF
RESOLUTION OF UNRESOLVED OPERATING REACTORS,
SAFETY ISSUE A-47 "SAFETY APPLICANTS FOR
IMPLICATION OF CONTROL OPERATING LICENSES
SYSTEMS IN LWR NUCLEAR AND HOLDERS OF
POWER PLANTS" PURSUANT TO CONSTRUCTION PERMITS
10 CFR 50.54(f) FOR LIGHT WATER
REACTOR NUCLEAR
POWER PLANTS
89-18 RESOLUTION OF UNRESOLVED 09/06/89 ALL HOLDERS OF
SAFETY ISSUE A-17, "SYSTEMS OPERATING LICENSES
INTERACTIONS IN NUCLEAR OR CONSTRUCTION
POWER PLANTS PERMITS FOR NUCLEAR
POWER PLANTS
- 2 -
propose changes to their TS that are consistent with the guidance in the enclosure.
Conforming amendment requests will be expeditiously reviewed by the NRC Project Manager for the facility. Please contact the Project Manager if you have questions on this matter.
Sincerely, Original signed by James G.Partlow James G. Partlow Associate Director for Projects Office of Nuclear Reactor Regulation Enclosures:
1. Guidance for the Modification of the Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications
2. List of Most Recently Issued NRC
Generic Letters Distribution:
See Attached Concurrence:
The following individuals concurred in this Generic Letter by their concurrence in the May 10, 1989 memorandum forwarding the Generic Letter to CRGR:
TGDunning, OTSB:DOEA; RLEmch, OTSB:DOEA; EJButcher, C:OTSB:DOEA;
BCalure, TECH EDITOR; CHBerlinger, C:GCB:DOEA; CERossi, D:DOEA:NRR;
FJCongel D:DREP:NRR; LCShao, D:DEST; SATreby, OGC; FJMiraglia, ADT:NRR;
JHSniezek, DD:NRR.
In a memorandum from ELJordan, Chairman, CRGR, to JHSniezek, Deputy Director, NRR, dated August 11, 1989, CRGR waived their review of this Generic Letter.
Jose A. Calvo, Chief Technical Specifications Branch Division of Operational Events Assessment, NRR
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